In a know type of nuclear reactor, for example as used in the Dresden Nuclear Power Station near Chicago, Ill., the reactor core comprises a plurality of fuel assemblies arranged in an array capable of self-sustained nuclear fission reaction. The core is contained in a pressure vessel wherein it is submerged in a working fluid, such as light water, which serves both as coolant and as a neutron moderator. A plurality of control rods, containing neutron absorbing material, are selectively insertable among the fuel assemblies to control the reactivity of the core. For further information on nuclear reactors see, for example, "Nuclear Power Engineering", M. M. El-Wakil, McGraw-Hill Book Company, Inc., 1962.
Each fuel assembly comprises a tubular flow channel, typically of approximately square cross section, containing an array of elongated, cladded fuel elements or rods supported between upper and lower tie plates. The fuel assemblies are supported in the pressure vessel between an upper core grid and a lower core support plate. The lower tie plate of each fuel assembly is formed with a nose piece which fits through an aperture in the core support plate into a pressurized coolant supply chamber. The nose piece is formed with openings through which the pressurized coolant flows upward through the fuel assembly flow channels to remove heat from the fuel elements. A typical fuel assembly of this type is shown, for example, by D. A. Venier et al. in U.S. Pat. No. 3,350,275. In nuclear reactors of recent design, in-core nuclear instrumentation, in the form of neutron detector, is contained in instrumentation receptacles located in the spaces or gaps between the fuel assemblies.
In a boiling water reactor, heat is transferred from the fuel through the fuel rod cladding to the water flowing upward among the fuel rods. At some elevation the flowing water reaches saturation temperature and beyond this point increasing fractions of the water are in the vapor phase. Normally the heat transfer coefficient between the fuel rod cladding and the water is substantially constant. However, if the heat-flux and consequently the steam fraction is increased sufficiently, a threshold is reached at which the heat transfer coefficient decreases suddenly by a factor of 5 to 10. This is caused by a change in the heat transfer mechanism from nucleate boiling to film boiling and it results in a very rapid, undesirable rise in fuel rod cladding temperature. The heat flux at the threshold between nucleate boiling and film boiling is designated the "critical heat flux."
An important consideration in the design of boiling water reactors is the relationship between the in-channel flow (or the coolant flow through the fuel assembly flow channels) and the bypass flow (or the coolant flow through the gaps among the fuel assemblies). On the one hand it is desirable to maximize the in-channel flow to thereby maximize the margin to critical heat flux. On the other hand it is necessary to provide a limited amount of bypass flow to avoid coolant stagnation and steam voids and to adequately cool the control rods and the in-core instrumentation devices located in the gaps between the fuel assembly flow channels. Thus for a given total core recirculation flow, the desired balance between in-channel and bypass flow maintains an adequate margin to critical heat flux while avoiding excessive out-of-channel voids.
In prior arrangments control of bypass flow is accomplished by allowing an amount of coolant leakage between the assembly flow channel and the lower tie plate. The flow channel is not fixed to the fuel assembly but is instead a slip fit over the upper and lower tie plates so that it readily can be removed during refueling and for inspection of the fuel rods and fuel assemblies. The flow channel is formed of relatively thin material to conserve space and to minimize parasitic neutron absorption and it is found that increases in pressure of the coolant (to increase coolant flow through the fuel assemblies) causes the flow channel to deflect and move away from the lower tie plate thus causing an excessive amount of bypass or leakage flow with the danger of depriving the fuel assembly of its required coolant flow.
Several arrangements have been proposed in the past to control excessive leakage flow created by such movement of the flow channel. Such arrangements are disclosed and claimed by B. A. Smith et al. in U.S. Pat. No. 3,689,358, C. R. Mefford et al. in U.S. Pat. No. 3,697,376 and by D. A. Venier et al. in U.S. Pat. No. 3,715,274. In these arrangements, see for example FIG. 2 of U.S. Pat. No. 3,689,358, the leakage flow is from the lower portion of the fuel assembly, downward between the lower tie plate and surrounding flow channel and out beneath the lower edge of the flow channel. Thus the leakage flow, together with the general coolant flow, has suffered the pressure drop across the fuel element support grid of the lower tie plate.
The object of the present invention is to provide an improved leakage flow control arrangement wherein the pressure difference between the coolant within the lower tie plate (below the fuel element support grid) and the bypass coolant outside the nose piece is utilized to maintain leakage flow control members in engagement with the surrounding flow channel.